1. Field of the Invention
This invention deals with materials and techniques for shielding of neutron and gamma radiation emitting together from radioactive waste sources such as transuranic and high-level wastes. It is based on specially formulated composite materials and techniques. In particular, this invention relates to different composite materials and admixtures, and their multifaceted application to safe handling, containerization and management of neutron and gamma emitting high-level, transuranic and low-level radioactive wastes and materials, as well as to decontamination and decommissioning of radioactively contaminated facilities. Owing to their significant capacity for attenuation of neutron and gamma radiation, these technologies relates to protecting health and environment from exposure to harmful radiation emitted by nuclear wastes and materials.
2. Description of the Related Art
Radioactive wastes, owing to temporal decay and fission of radionuclides, emit alpha, beta, gamma and neutron radiation, of which neutron and gamma radiation are extremely harmful. Radioactive wastes can be solids, liquids and sludge, and these are of three types:
A) High-level radioactive wastes contain gamma emitting long-half life radionuclides, such as plutonium (Pu-238, Pu-239, Pu-240 and Pu-242) and uranium (U-234, U-235, and U-236). High-level wastes include spent (or used up) nuclear fuel and wastes from commercial and defense related nuclear reactors resulting from reprocessing of spent nuclear fuel. Most spent nuclear fuel in the United States is currently located in pools of water, at nuclear generating plants across the country, to protect workers from radiation. Spent fuel also is stored in large concrete casks. High-level wastes are also generated from reprocessing of fuel from weapons production reactors to obtain materials to make nuclear weapons. These wastes are primarily in liquid and sludge forms.                B) Transuranic (TRU) wastes contain such radionuclides as californium Cf-249-252), americium (Am-241, 242 and 243), curium (Cm-242-250), neptunium (Np-235 and 236), plutonium (Pu-236-242) and berkelium (Bk-247 and 250). Generally, TRU wastes are made up of solids or liquids and contain radionuclides that have more than 20-year half-lives. TRU wastes are generated by defense nuclear research and development activities, such as development and fabrication of nuclear weapons. TRU wastes are usually classified as “contact-handled” (CH) and “remote-handled” (RH) wastes. These are highly radioactive with high radiation flux of neutrons and gamma rays, as well as alpha rays. Often, these wastes are mixed with hazardous organic and inorganic wastes, and therefore, they are also called as transuranic mixed wastes.        
C) Low-Level radioactive wastes do not include either high-level or transuranic waste materials. Most low-level wastes (classified by the NRC as A, B or C) emit relatively low-levels of radiation from radioactive decay of short half-life radionuclides, such as strontium-90, cesium-137, krypton-85, barium-133 and beryllium-7 and 10. Generally, these wastes have radioactivity that decays to background levels in less than 500 years and about 95 percent of the waste decays to background levels in about 100 years. Low-level radioactive wastes are generated by commercial and university laboratories, pharmaceutical industries and hospitals, as well as nuclear power plants. Low-level wastes include both solid and liquid wastes.
High-level wastes are very radioactive, which emit extremely harmful gamma (like x-rays) and neutron radiation. RH-TRU wastes are primarily neutron and secondary gamma radiation emitters, CH-TRU wastes are also very radioactive, which emit harmful alpha radiation, as well as neutron radiation. In order to handle these wastes, heavy concrete and/or lead shielding materials are required and high energy flux energy radioactive wastes, such as RH-TRU wastes, are robotically handled despite the concrete/lead shielding. One of the main radiation hazards posed by this waste is through exposure and inhalation or ingestion. During handling and management, inhalation of or exposure to certain transuranic wastes, such as plutonium in very small quantities, could deliver significant internal radiation doses.
Exposure to gamma and neutron radiation, as well as alpha and beta radiation, associated with these wastes can induce chronic, carcinogenic and mutagenic health effects that lead to cancer, birth defects and death. However, thousands of tons of both solid and liquid, as well as sludge radioactive wastes have been generated in the past and they will continue to be generated in the future by commercial/private industries and government agencies. Unless they are safely and cost-effectively shielded, managed and disposed, these wastes may pose serious health and economic consequences.
Generally, alpha radiation can be easily shielded by paper, skin or clothes, where as beta radiation can easily pass through paper, skin or clothes but it will be blocked by a thin layer of plastic, aluminum foil or wood. In contrast, gamma and neutron radiation is very penetrating, and neutron radiation is more penetrating than gamma. Gamma radiation can be blocked by heavy shielding materials such as thick-concrete, lead, steel and Ducrete (depleted uranium mixed with concrete); whereas neutron radiation can penetrate through heavy metal shielding, only specially engineered and chemically formulated high density concrete blocks and lead can shield penetration of neutron radiation from its source.
High-level radioactive wastes are currently stored at nuclear power plants and DOE facilities across the country. Similar wastes have been generated by the Department of Defense also. Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) is charged with identifying and developing a suitable site for deep geologic disposal of these wastes. The OCRWM is currently conducting research and testing to determine the suitability of the Yucca Mountain, Nev. site for long-term safe disposal of these wastes. Transuranic wastes are destined to be disposed into an already established geologic repository at WIPP site in Carlsbad, N. Mex. Class A and B low-level radioactive wastes are currently disposed in isolated shallow burial ground; whereas greater than class C waste low-level waste use deep geologic disposal in specially licensed facilities.
Management and disposal of high-level, transuranic and low-level radioactive wastes are very risky. Radioactive waste management also includes decontamination and decommissioning of contaminated sites. Management activities, prior to disposal, include handling, solidification of liquid wastes, loading, storage, radiation monitoring, reloading of wastes into transportable containers, and transport of waste containers to long-time safe disposal sites. Storage, transportation and disposal of radioactive wastes are a growing problem in the United States and abroad. Many U.S. commercial power plants do not have sufficient existing capacity to accommodate future spent nuclear fuel wastes, and much of the DOE's HLW and TRU wastes are currently located in unlicensed storage structures that need to be upgraded or replaced. Therefore, there is a strong need for improved radiation shielding materials and techniques for waste container systems so that the wastes can be safely stored, transported and disposed.
Currently, two main methods are used for storage of commercial power plant nuclear waste: wet and dry. In wet storage, the waste is immersed in a lined, water-filled pool, which shields the radiation and removes radioactive heat aided by an active system. Wet storage is intended for a period of five years after waste immersion, and thereafter, it is stored in dry storage casks or vaults constructed out of concrete, which shield the radiation. Generally, the design and manufacturing of waste containment systems for the dry storage are governed by a number of governing factors, such as 1) shielding effectiveness, 2) structural integrity and durability, 3) thermal performance, 4) ease of handling and transportation, 5) high volume waste loading, 6) cost-effectiveness, and 7) health and environmental protection.
Current radiation shielding and waste containment technologies are based on low or high density concrete, lead, carbon and stainless steel, borated resins, polymers and other additives, as well as glass vitrification and ceramic calcinations. However, these materials and processes have limitations and they do not fully satisfy the above-mentioned governing factors of waste containment systems. Some examples of these limitations are as follows:                The above mentioned shielding materials or additives and technologies do not meet the shielding requirements of radiation waste sources consisting of a flux of mixed radiation types of various energy levels and the secondary radiation effects (e.g., emission of secondary gamma radiation due to inelastic collision or capture of emitted neutrons) that are induced within the shields as a result of interaction of the initial flux with certain atoms in the shield itself.        While thin liners of lead, used in waste storage casks and containers, are effective for shielding gamma radiation, they are not very effective in shielding neutron radiation. When applied as a part of a neutron particle shielding, lead has an extremely low level of neutron absorption, and hence, practically no absorption of secondary gamma radiation. For neutron shielding, thicker lead liners are required, which not only reduces the space for waste loading in the containment systems but also makes the containment systems heavy for handling and transport. Consequently, lead technology can be costly. If the shield material has a high rate of neutron capture, it will over time become radioactive, and sharply reduce its effectiveness as a shield material, consequently, their subsequent handling and disposal will be a problem. In addition, lead can be leached and will contaminate the environment, potentially posing toxic health effects.        Although some containment systems have used concrete liners, castings or grouts as safe storage of radioactive wastes, they are not very effective in shielding high energy flux of neutron and gamma radiation, unless significantly thick high density concrete liners in conjunction with metal liners are used. Generally, concrete liners are not very efficient in shielding neutron radiation because, concrete products have low hydrogen atomic density, which is the measure of a materials ability to shield neutron radiation. In addition, concrete-based containment systems generally lack mobility, and therefore, limit the volume of radioactive wastes that can be stored in a given limited space due to the high density and volume concrete required to obtain the necessary shielding properties. As a result, the application of this technology to waste containment systems can be uneconomical. In addition, chemical and mechanical properties of concrete can be degraded due to alkali-silica-reaction (at <5 pH) and at elevated radioactive temperatures, resulting in shrinkage and cracking and consequential attenuation of its shielding capacity. Similarly, the bonded water in cement grouts tends to decrease with time due to radioactive heat, causing increase in porosity and reduction in shielding capacity. Traditionally, Portland cement-based grouts have been used for solidification/encapsulation of hazardous and low level radioactive wastes. However, this technology has shown to be effective only in situations where the salt loading is relatively low (i.e. <10%) and when the total organic content of the waste is below 3%. Given the above limitations, use of concrete based technology for solidification of liquid wastes and storage of high-level and transuranic wastes may be inappropriate.        Borated stainless steel has been used in the radioactive waste storage containers; however, this material, owing to its weak mechanical/metallurgical properties, has the potential for cracking and breaking, rendering weak shielding capacity over a long period of time. Further, the bombardment of borated stainless steel by the neutrons emitted by the wastes can reduce the steel's shielding efficacy, making it an unsuitable material for long term safe storage of high-level and transuranic wastes.        In the case of vitrification technology, there is significant uncertainty in effectiveness of in-situ or ex situ vitrification technology for solidification of liquid wastes with variable compositions and pH conditions, as well as for volatile components. In addition, glass production and chemical durability of vitrified glass is unknown. In glass production, the largest uncertainties are related to the reliability and safety of the high-temperature melting process behavior of the glass during the first and second glass pours, such as the effects of glass fracturing on chemical and physical durability, and the significance of mixed waste-constituents crystallization. Owing to rapid cooling rate and high viscosity of oxide and silicate, waste constituents/molecules cannot move sufficiently to be uniformly incorporated into crystalline structure of the glass. Furthermore, vitrification may produce secondary wastes and management of such wastes would be an issue to contend with. In terms of chemical durability of glass, very little is known about the type and conditions of formation of colloids and less about their ability to bind up and transport the waste constituents. Corrosion of vitrification melt materials from acidic wastes is a key issue that must be dealt with.        
In an attempt to reducing the thickness of concrete shield while maintaining the desired long-life of the waste containers, Suzuki et al (U.S. Pat. No. 4,687,614) taught a three layered structure comprising a metallic vessel with a reinforced concrete lining as an inner layer, and polymerized and cured impregnated layer as intermediate layer between the inner concrete layer and the outer metallic layer. However, this and similar other attempts have been unsuccessful in achieving the desired reduction in thickness. In addition, this three layered system was found to be not very effective in shielding high energy flux of neutron and gamma radiation.
Kronberg (U.S. Pat. No. 5,334,847) teaches an alternate shielding system using depleted uranium core for absorbing gamma rays with a bismuth coating for preventing corrosion, and alternatively having a gadolinium sheet positioned between the depleted uranium core and the bismuth coating for absorbing neutrons. However, this shielding system does not reduce the undesirable density and thickness of the shielding to maintain the desired capacity for shielding of high flux neutron and gamma radiation. In addition, this shielding system is neither efficient in avoiding the depleted uranium corrosion nor assuring the durability of the shielding system over desired long-life, particularly at elevated temperatures. Owing to the uranium corrosion, this system is considered inefficient for shielding of neutron and gamma radiation fluxes. In addition, corrosion can cause leaching and release of uranium from the concrete in the form of uranium bicarbonate and uranium tri-carbonate complexes, causing health and environmental problems. Furthermore, this system is relatively expensive.
Yoshihisa, in Japanese Patent Document No. 61-091598, teaches utilization of depleted uranium and uranium oxide aggregate containing concrete for radiation shielding. While this system has the potential for reducing the thickness of radiation shielding for gamma rays, it has serious problems of concrete degradation and maintaining the desired long-life of the system, particularly at elevated radioactive temperatures. Tensile and compressive strengths of concrete are seriously compromised by addition of the uranium aggregate to the concrete. Quapp et al. (U.S. Pat. Nos. 5,786,611 and 6,166,390) disclose radiation shielding of containers for storing spent nuclear fuel waste. These containers are formed from concrete product with stable uranium oxide aggregate and a neutron absorbing material. The neutron absorbing materials described are B2O3, HfO2 and Gd2O3. In addition, the concrete shielding composition of this invention requires including reinforcing materials. These may include, steel bars, fillers and strengthening impregnates, such as steel fiber, glass fiber, polymer fiber, lath or steel mesh, creating a complex system of shielding.
However, owing to the uranium corrosion problem, this concrete shielding products along with their additives are not efficient for radiation shielding and they do not contribute to the long-time durability of waste containers, especially at elevated temperature. Corrosion can cause leaching and release of uranium from the concrete in the form of uranium bicarbonate and uranium tricarbonate complexes, causing health and environmental problems. Further, this type of shielding containers does not reduce the undesirable density and thickness of the shielding to maintain the desired capacity for shielding of high flux neutron and gamma radiation. In addition, cooling of concrete surfaces is required during radioactive waste storage to further the length of the concrete to avoid high radioactive temperature, without which, the concrete system could degrade and allow for emission of radiation. Generally, concrete systems lack mobility and limit the volume of radioactive wastes to be stored in a given space due to great concrete thickness and density required to obtain the necessary shielding properties.
The above mentioned shielding materials and systems, using single component or dual component materials provide only limited shielding capacity under a given set of density, thickness and configuration of shielding materials and containers. Generally, they do not offer the desired shielding of both neutron and gamma emitted from the same waste source, particularly the transuranic waste source or its containers. These materials and techniques suffer from the problems of offering desired shielding efficiency, long-term durability, health and environmentally safety. In addition, the systems are complex and made up of multilayered dense and thick layers of concrete admixed with depleted uranium, lead and stainless steel, which reduce the volume of containers/casks for radioactive waste loading. Consequently, more containers/casks have to be built to store or transport a given volume of radioactive wastes; therefore, those containment systems are not cost-effective. Furthermore, high density containment systems are not be easily mobile and are very difficult to handle, in addition to being unsafe.
In general, the prior art uses many kinds of additives to meet the shielding requirements of a particular radiation spectrum and energy flux involved, but they are not effective in meeting the desired shielding requirements of radiation fluxes of different energy levels arising from complex, uncharacterized radioactive waste sources. This situation may be further complicated when secondary radiation effects are induced as a result of interaction of initial radiation flux with certain atoms in the waste materials, as well as within a given shielding material. Therefore, it is necessary to formulate admixture composite materials that offer optimal total radiation shielding capacity to cater to the needs of such complexities.
Accordingly, it is desirable and advantageous to provide improved materials and simple techniques that offer a better, more durable and cost-effective radiation shielding and waste containment systems than those mentioned above. Improved materials and techniques shall enhance the safety of handling, storage, transportation, long-time containment of radioactive wastes, as well as protect human health and environment. In addition, it is desirable for such materials and techniques to have such attributes as a) applicable to shield multi spectral and energy flux radiation, b) ease of application, c) easy to handle variations in waste characteristics without the need for separation of incompatible wastes that do not generate secondary waste streams, d) will not expose workers to any significant and unnecessary amount of radiation and e) exhibit superior performance over regulatory long times.